期刊文献+

Inconel 690合金在含Pb碱溶液中的应力腐蚀行为 被引量:2

Effect of heat treatment on stress corrosion behavior of alloy 690 in caustic solution containing Pb
在线阅读 下载PDF
导出
摘要 采用扫描电镜(SEM)、背散射电子衍射分析技术(EBSD)等方法分析不同温度固溶处理(1090℃,1100℃)的Inconel690合金微观组织,并用恒变形方法研究了该Inconel 690合金在含Pb氢氧化钠溶液中的应力腐蚀行为。结果表明,固溶处理温度1100℃的试验合金具有更好的组织均匀性,低重合点阵晶界所占比例更高。固溶处理温度1090℃的试验合金在含Pb氢氧化钠溶液中具有更强的应力腐蚀敏感性。综合分析试验结果,较低的固溶处理温度导致690合金的低重合点阵晶界所占比例低,增加了其在含Pb氢氧化钠溶液中的应力腐蚀敏感性。 Microstructure of alloy 690 solid-solutioned at different temperatures was investigated by SEM and EBSD technique. The stress corrosion behavior of alloy 690 in NaOH solution containing Pb was studied by a constant strain method. SEM and EBSD results show that the microstructure of alloy 690 solutioned at 1100 ℃ is more uniform. Alloy 690 mill treated at 1100 ℃ has the higher proportion of low coincidence site lattice grain boundaries. The results of stress corrosion test at high temperature and high pressure condition show that alloy 690 treated at 1090℃ exhibits higher stress corrosion sensitivity in the test solution. The alloy 690 solution-treated at the lower temperature has the lower proportion of low coincidence site lattice grain boundaries,which increases the stress corrosion sensitivity in NaOH solution containing Pb.
出处 《材料热处理学报》 EI CAS CSCD 北大核心 2013年第10期38-42,共5页 Transactions of Materials and Heat Treatment
基金 国家自然科学基金(钢铁联合研究基金)(U1260201) 国家高技术研究发展计划(863)(2012AA050901)
关键词 690合金 热处理 SEM 应力腐蚀 alloy 690 heat treatment SEM stress corrosion lead
作者简介 李成涛(1980-),男,高级工程师,从事核电材料的腐蚀性能研究,发表论文10余篇,电话:0512-68601473,E-mail:lctslj0630@hotmail.com 通讯作者:李晓刚(1963~),男,教授,电话:010-62333975,E-mail:lixiaogang99@263.net。
  • 相关文献

参考文献16

  • 1Copson H R, Dean S W. Effect of contaminants on resistance to stress corrosion cracking of Ni-Cr alloy 600 in pressurized water[ J]. Corrosion, 1965, 21(1): 1-8.
  • 2Durra R S,Tewari R, De P K. Effects of heat-treatment on the extent of chromium depletion and caustic corrosion resistance of alloy 690 original research [ J ]. Corrosion Science,2007,49 ( 2 ) : 303 - 318.
  • 3Casales M, Salinas-Bravo V M, Martinea-Villafane A, et al. Effect of heat treatment on the stress corrosion cracking of alloy 690 [ J]. Materials Science and Engineering A,2002,332 ( 1 - 2 ) : 223 - 230.
  • 4Dutta R S, Tewafi R, De P K. Effects of heat-treatment on the extent of chromium depletion and caustic corrosion resistance of alloy 690 [ J]. Corrosion Science,2007,49 (2) :303 - 318.
  • 5Was G S, Rajan V B. Mechanism of intergranular cracking of Ni-Cr-Fe alloys in sodium tetrathionate [ J ]. M etall Trans A, 1987,18 (7) :1313 -1323.
  • 6Was G S. Grain boundary chemistry and intergranular fracture in austenitic nickel-base alloys[ J]. Materials Science Forum, 1989,46:335 -358.
  • 7Kim U C,Kim K M, Lee E H. Effects of chemical compounds on the stress corrosion cracking of steam generator tubing materials in a caustic solution [ J ]. Journal of Nuclear Materials ,2005,341 ( 2 ) : 169 - 174.
  • 8Peng B,Lu B T,Luo J L,et al. Investigation of passive films on nickel Alloy 690 in lead-containing environments[ J ]. Journal of Nuclear Materials, 2008,378(3) :333 - 340.
  • 9李成涛,刘飞华,李岩,程学群,董超芳,李晓刚.固溶温度对690合金在NaOH溶液中的电化学性能影响[J].材料热处理学报,2011,32(9):48-52. 被引量:2
  • 10华惠中,黄春波,吕战鹏,杨武.800、600和690合金的铅致应力腐蚀破裂[J].腐蚀与防护,2001,22(11):483-488. 被引量:13

二级参考文献46

  • 1刘小萍,田文怀,杨峰,郭朝丽,高原.时效处理SUS316L不锈钢中析出相的晶体结构和化学成分[J].材料热处理学报,2006,27(3):81-85. 被引量:15
  • 2[1]Sarver J M. Information on lead concentration for BWI water chemistry manual[R]. Internal Report, B&W25211-002,1994.
  • 3[2]Agrawal A K, Paine J P N. Lead cracking of alloy 600-A review[A]. Proc 4th Iht Symp on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Jekyll Insland GA:1989.
  • 4[3]Miglin B P, Sarver J M. Investigation of lead as a cause of stress corrosion at support plate interatctions[R]. EPRI NP-7367-S, 1991.
  • 5[4]Sakai T,Senjuh T, Aoki K, et al. Lead-induced stress corrosion cracking of alloy 600 and 690 in high temperature water[A]. Proc 5th Int Symp on Environmental Degradation of Materials in Nuclear Power SystemsWater Reactors[C]. Monterey, CA:1991,764.
  • 6[5]Miglin B P, Sarver J M, Pasaila-Dombrowski M J, et al. Lead assisted stress corrosion crocking of nuclear steam generator tubing materials[A]. Proc of Improving the Understanding and Control of Corrosion on the Secondary Side of Steam Generators[C]. Airlie,VA: 1995,95.
  • 7[6]Hayner G, Frye C, Theus G, et al. Examination of tubes removed from st. lucie unit 1 and investigation of causes of the corrosion[A]. Proc of the Third International Symposium on Environment Degradation of Materials in Nuclear Power Systems-Water Reactors[C].TMS, Traverse City, MI: 1987. 449.
  • 8[7]King P J, Gonzalez F, Brown J. Stress corrosion cracking experience in steam generators at bruce NGS[A]. Proc of the 62th Iht Symp on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors[C]. San Diego, CA: 1993. (Minerals Metals Materials Society, Warrendale, PA) 233.
  • 9[8]Rocher A,Nordmann F. Transport of lead in secondary system of PWR plarts: laboratory and plant investigations[A]. 6th Conf on Water Chemistry of Nuclear Reactor System, Bournemouth [C]. British Nuclear Energy Soc. , London: 1992. 249.
  • 10[9]Agrawal A L Lead cracking of alloy 600[R]. Paper Presented at EPRI Meeting on Lead SCC of Alloy 600,Charlotte NC.. 1988.

共引文献13

同被引文献16

  • 1Hui Li,Shuang Xia,Bangxin Zhou,Wenqing Liu.C–Cr segregation at grain boundary before the carbide nucleation in Alloy 690[J]. Materials Characterization . 2012
  • 2Q.J. Peng,J. Hou,T. Yonezawa,T. Shoji,Z.M. Zhang,F. Huang,E.-H. Han,W. Ke.Environmentally assisted crack growth in one-dimensionally cold worked Alloy 690TT in primary water[J]. Corrosion Science . 2012
  • 3Z.M. Zhang,J.Q. Wang,E.-H. Han,W. Ke.Trans-twins stress corrosion cracking behaviors of Alloy 690TT in lead-containing caustic solution at 330<ce:hsp sp="0.25"/>°C[J]. Nuclear Engineering and Design . 2011 (12)
  • 4Shuang Xia,Bangxin Zhou,Wenjue Chen.Grain Cluster Microstructure and Grain Boundary Character Distribution in Alloy 690[J]. Metallurgical and Materials Transactions A . 2009 (12)
  • 5B. Peng,B.T. Lu,J.L. Luo,Y.C. Lu,H.Y. Ma.Investigation of passive films on nickel Alloy 690 in lead-containing environments[J]. Journal of Nuclear Materials . 2008 (3)
  • 6R.S. Dutta,R. Tewari,P.K. De.Effects of heat-treatment on the extent of chromium depletion and caustic corrosion resistance of Alloy 690[J]. Corrosion Science . 2006 (2)
  • 7Bogdan Alexandreanu,Gary S. Was.The role of stress in the efficacy of coincident site lattice boundaries in improving creep and stress corrosion cracking[J]. Scripta Materialia . 2005 (6)
  • 8B. Ravi Kumar,Raghuvir Singh,Bhupeshwar Mahato,P.K. De,N.R. Bandyopadhyay,D.K. Bhattacharya.Effect of texture on corrosion behavior of AISI 304L stainless steel[J]. Materials Characterization . 2004 (2)
  • 9Yun Soo Lim,Joung Soo Kim,Hong Pyo Kim,Hai Dong Cho.The effect of grain boundary misorientation on the intergranular M 23 C 6 carbide precipitation in thermally treated Alloy 690[J]. Journal of Nuclear Materials . 2004 (1)
  • 10M Shimada,H Kokawa,Z.J Wang,Y.S Sato,I Karibe.Optimization of grain boundary character distribution for intergranular corrosion resistant 304 stainless steel by twin-induced grain boundary engineering[J]. Acta Materialia . 2002 (9)

引证文献2

二级引证文献5

相关作者

内容加载中请稍等...

相关机构

内容加载中请稍等...

相关主题

内容加载中请稍等...

浏览历史

内容加载中请稍等...
;
使用帮助 返回顶部